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Catalog of Programs in Category C
C. Static Design Studies
| nesc0325 | 2-DB, 2-D MultiGroup Diffusion, X-Y, R-Theta, Hexagonal Geometry Fast Reactor, Criticality Search |
| nea-0108 | ALCI, Homogeneous 2-D MultiGroup Neutron Diffusion in X-Z, R-Z, R-Theta Geometry with Criticality Search |
| ccc-0558 | ALKASYS, Rankine-Cycle Space Nuclear Power System |
| psr-0315 | AMPX-77, Modular System for Coupled Neutron-Gamma MultiGroup Cross-Sections from ENDF/B-5 |
| nea-0633 | ANIPLO-D50, Plot of Scalar Flux and Dose Rates from ANISN Calculation |
| ccc-0082 | ANISN-E, 1-D Transport Program ANISN with Exponential Model |
| nea-0363 | ANISN-FONTENAY, 1-D Planar, Spherical, Cylindrical Neutron Transport and Gamma Transport with Deep Penetration |
| ccc-0082 | ANISN-JR, 1-D Transport Program ANISN with ZZ JSD Data and Flux Plot |
| ccc-0254 | ANISN-ORNL, 1-D Neutron Transport & Gamma Transport in Slab, Cylindrical, Spherical Geometry with Anisotropic Scattering |
| ccc-0255 | ANISN-W, 1-D Transport Calculation for Deep Penetration Problems |
| ccc-0514 | ANISN/PC, MultiGroup 1-D Discrete Ordinates Transport with Anisotropic Scattering |
| nea-0546 | APPLE, Plot of 1-D MultiGroup Neutron Flux and Gamma Flux and Reaction Rates from ANISN |
| nea-0320 | ARGO, 1-D Neutron Diffusion in Slab, Cylindrical, Spherical Geometry from JAERI Fast-Set, ABBN, RCBN |
| nea-0661 | ASNBILD, Generator of JCL and Data for Program ANISN on CDC Computer |
| ccc-0519 | AUS, Neutron Transport and Gamma Transport System for Fission Reactors and Fusion Reactors |
| nea-0179 | AXIFLUX, Cosine Function Fit of Experimental Axial Flux in Cylindrical Reactor |
| ccc-0459 | BOLD/VENTURE-4, Reactor Analysis System with Sensitivity and Burnup |
| nea-1187 | BOREAS, Nuclear and Thermohydraulic LWR Burnup Simulation |
| nea-1678 | BOT3P5.3, 3D Mesh Generator and Graphical Display of Geometry for Radiation Transport Codes, Display of Results |
| nea-1523 | BOXER, Fine-flux cross section condensation, 2D few group diffusion and transport burnup calculations |
| iaea1190 | BRETISLAV,OLDRICH, Neutron Diffusion in Hexagonal Geometry for WWER Critical Fuel Assemblies |
| nesc0270 | CAESAR-4, 1-D MultiGroup Diffusion in Slab, Cylindrical, Slab Geometry, Criticality Search |
| nea-0649 | CARMEN-SYSTEM, Programs System for Thermal Neutron Diffusion and Burnup with Feedback |
| iaea0920 | CEBIS, 1-D 2 Group Diffusion Code for Reactor Calculation |
| nesc0387 | CITATION, 3-D MultiGroup Diffusion with 1st Order Perturbation and Criticality Search |
| ccc-0643 | CITATION-LDI2, 2-D MultiGroup Diffusion, Perturbation, Criticality Search, for PC |
| nea-0357 | CLUP-77C, Collision Probability and Neutron Flux Distribution in Multi-Region Fuel Cluster |
| ccc-0726 | CNCSN 2009, One, Two- and Three-Dimensional Coupled Neutral and Charged Particle Sn Parallel Multi-Threaded Code System |
| ccc-0777 | COG 11, Multiparticle Monte Carlo Code System for Shielding and Criticality Use |
| iaea1226 | CORD, PWR Core Design and Fuel Management |
| nea-0057 | CRAM-360, 1-D, 2-D Multi Group Diffusion with Keff Calculation or Criticality Search |
| nea-1416 | D3D/D3E, 2-D, 3-D MultiGroup Neutron Diffusion in Rectangular, Cylindrical, Triangular-Z Geometry |
| nea-0672 | DIAMANT-2, MultiGroup Neutron Transport with Anisotropic Scattering in Triangular Geometry |
| ccc-0649 | DIF3D 8.0/VARIANT8.0, 2-D 3-D Multigroup Diffusion/Transport Theory Nodal & Finite Difference Solver, Variational Method |
| ccc-0784 | DIF3D10.0, Variational Nodal Methods, Finite Difference Methods to Solve N diffusion & Transport Theory Problems |
| nea-0808 | DIFFUSION-ACE, 3-D Neutron Diffusion by Leakage Iteration Method |
| nea-0184 | DIXY-2, 2-D Homogeneous and Inhomogeneous Neutron Diffusion N X-Z, R-Z, R-Theta Geometry with Perturbation |
| nea-0391 | DLS, 2-D Diffusion with Line-of-Sight Method for Cavities |
| iaea1241 | DNTM/R2D, 2-D Transport in X-Y Geometry |
| ccc-0650 | DOORS, 1-,2-,3-dimensional discrete-ordinates system for deep-penetration neutron and photon transport |
| ccc-0543 | DORT, 1-D 2-D Discrete Ordinate Neutron and Photon Transport with Deep Penetration |
| ccc-0276 | DOT-3.5, 2-D Neutron Transport, Gamma Transport Program DOT with New Space-Scaling |
| ccc-0320 | DOT-4.2, 2-D Neutron Transport, Gamma Transport with Space Dependent Mesh and Quadrature |
| nea-1506 | DPOL3D, 2 Group, 3-D Core Transients and Steady State |
| uscd1234 | DRAGON 3.05D, Reactor Cell Calculation System with Burnup |
| ccc-0647 | DRAGON, Reactor Cell Calculation System with Burnup |
| nesc0209 | DTF-4, 1-D MultiGroup Time-Independent Boltzmann Equation, Slab, Cylindrical, Spherical Geometry, Sn-Method, Pl-Method |
| nea-0269 | DTF-G, Reactivity and Flux by 1-D Sn Method on Planar Cylindrical Spherical Geometry |
| nea-0322 | DTF4-J, 1-D Neutron Transport with Anisotropic Scattering by Sn Method |
| nea-1683 | ERANOS 2.0, Modular code and data system for fast reactor neutronics analyses |
| nea-0534 | EREBUS, Burnup by 2-D MultiGroup Neutron Diffusion with Criticality Search |
| nea-0449 | ESTOQ, 1st Collision Source Calculation for Program DOT in R-Z Geometry |
| nea-0311 | EXPANDA-4, 1-D Neutron Diffusion in Slab, Spherical Geometry from JAERI Fast-Set with Criticality Calculation |
| nea-0312 | EXPANDA-5, 1-D Neutron Diffusion in Slab, Cylindrical, Spherical Geometry for 2 Region Fast Reactor Criticality |
| nea-0313 | EXPANDA-6, 1-D Neutron Diffusion in Slab, Cylindrical, Spherical Geometry from JAERI Fast-Set with PuO2 UO2 Mixture |
| nea-0315 | EXPANDA-70, 1-D Neutron Diffusion in Slab, Cylindrical, Spherical Geometry with Criticality Search |
| nesc0156 | EXTERMINATOR-2, 2-D MultiGroup Neutron Diffusion in X-Y R-Z or R-Theta Geometry |
| iaea0835 | FASVER, 2 Group 2-D Diffusion in X-Y Geometry and Adjoint Solution for PWR with Reflector |
| psr-0563 | FEAST-METAL-V.1.0, Fuel Engineering and Structural analysis Tool |
| nea-0443 | FEM-2D, 2-D MultiGroup Diffusion in X-Y Geometry |
| nea-0545 | FEM-BABEL, 3-D MultiGroup Neutron Diffusion by Galerkin Method |
| nea-0566 | FEM-RZ, 2-D MultiGroup Neutron Transport in R-Z Geometry, Eigenvalue and Fixed Source Problems |
| nea-0478 | FEMB, 2-D Homogeneous Neutron Diffusion in X-Y Geometry with Keff Calculation, Dyadic Fission Matrix |
| ests0486 | FESH, 2-D Multigroup Neutron Transport with Isotropic Scattering |
| iaea1221 | FIGA, Source Distribution Conversion from X-Y to R-Theta Geometry |
| nea-0896 | FINELM, MultiGroup Diffusion in 3-D by Finite Elements Method |
| nesc0167 | FLARE, 3-D BWR Reactivity and Power Distribution Appraisal Calculation |
| nea-0596 | FOCUS, Neutron Transport System for Complex Geometry Reactor Core and Shielding Problems by Monte-Carlo |
| nesc0028 | FOG-1-2-3, 1-D Few-Group Diffusion in Slab Cylindrical Spherical Geometry, Criticality Search, Buckling |
| ccc-0603 | FPZD, Reactor Burnup by MultiGroup Neutron Diffusion |
| nea-1021 | FURNACE, Neutronic Calculation in 3-D Toroidal Geometry |
| nesc0606 | GAPER-1D, 1-D MultiGroup 1st Order Perturbation Transport Theory for Reactivity Coefficient |
| nesc0380 | GATT, 3-D Few-Group Neutron Diffusion for Power Distribution in Hexagonal Reactor Core for HTGR |
| nea-0605 | GENP-2, Program System for Integral Reactor Perturbation |
| psr-0304 | GIRAFFE, Isotope Release Analysis in LMFBR Fuel Elements Failure |
| iaea1271 | GNOMER, Core Power Distribution by 1-D, 2-D, 3-D MultiGroup Neutron Diffusion |
| iaea0908 | GRENADE, Green's Function Nodal Algorithm for Diffusion Equation |
| ccc-0276 | GRUNCLE, 1st Collision Source Calculation for Program DOT |
| nesc0277 | HAMMER, 1-D MultiGroup Neutron Transport Infinite System Cell Calculation for Few-Group Diffusion Calculation |
| nesc0136 | HERESY, 2-D Few-Group Static Eigenvalues Calculation for Thermal Reactor |
| nea-0176 | HETERO, Flux and Power Distribution in Thermal Reactor by 3-D, 2 Group Line Source Sink Method |
| iaea1240 | HEXAB-3D, 3-D Few-Group Diffusion for Hexagonal Core Geometry |
| iaea0914 | HEXNOD23, 2-D, 3-D Coarse Mesh Solution of Steady State Diffusion Equation in Hexagonal Geometry |
| nea-1871 | JN-METD, N Transport with Isotropic Scattering, Bare Slabs and Homogeneous Slabs (JN-METD1), Multilayer Slabs (JN-METD2) |
| nea-0624 | JOSHUA, Neutronics, Hydraulics, Burnup, Refuelling of LWR |
| nea-0343 | KASY, 3-D Homogeneous Neutron Diffusion in X-Y-Z, R-Theta, Hexagonal-Z Geometry by Synthesis Method |
| ccc-0510 | KENO-4(RG), KENO-4 with Random Geometry |
| ccc-0436 | KENO-4/CRC, MultiGroup 3-D P1 Scattering Monte-Carlo Transport Calculation with Neutron Balance Edit |
| nea-1467 | KENO-VA-PVM KENO-VA-SM, KENO5A for Parallel Processors |
| ccc-0548 | KENO5A-PC, Monte-Carlo Criticality with Supergrouping |
| nea-0616 | KIM, Steady-State Transport for Fixed Source in 2-D Thermal Reactor by Monte-Carlo |
| iaea1232 | LABAN-PEL, 2-D MultiGroup Neutron Diffusion in X-Y Geometry by Response Matrix Eigenvalue |
| nea-0167 | LIE-PN, Pn Neutron Transport in Radial Geometry Cell with Source Problems Calculation |
| nea-0836 | MADONNA, Neutron Flux with Void Region by Removal Diffusion Method |
| nea-0528 | MARC, MultiGroup Diffusion in R-Z, X-Y, R-Theta, Slab, Cylindrical, Spherical, Hexagonal, Triangular Geometry |
| nea-0926 | MARIA-SYSTEM, PWR Fuel Assembly Calculation for Program CARMEN-System and Simulation |
| nea-1643 | MCB1C, Monte-Carlo Continuous Energy Burnup Code |
| nea-1733 | MCNP4B-GN, Monte Carlo Code System for (gamma,n) production and transport in high-Z materials |
| iaea0889 | MCRAC, In Core Fuel Management, Program of PFMP System |
| nea-1005 | MOBIDIC, Fast Reactor Hexagonal Infinite Lattice 2 Component Fuel Pin Diffusion Coefficient |
| iaea1238 | MOCA, Criticality of VVER Reactor Hexagonal Fuel Assemblies |
| nea-1279 | MODICO, 1-D Time-Dependent 1 Group, 2 Group Neutron Diffusion with Delayed Neutron Precursors |
| nea-0527 | MONK, Keff, Collision Rate, Flux Distribution in General Geometry from UKNDL by Monte-Carlo Method |
| nea-1747 | MONTE-CARLO-WS-2005, Proceedings of Monte Carlo Criticality Calculations & TRIPOLI-IV Workshops 2005 |
| ccc-0127 | MORSE, MultiGroup Neutron Transport and Gamma Transport for Complex Geometry Shields by Monte-Carlo |
| ccc-0431 | MORSE-C, Neutron Transport, Gamma Transport for Criticality Calculation by Monte-Carlo Method |
| ccc-0474 | MORSE-CGA, Monte-Carlo Neutron and Gamma MultiGroup Transport with Array Geometry |
| nea-1181 | MORSE-DD, Monte-Carlo Neutron Transport with Combinatorial Geometry Using DDL Cross-Sections Library |
| ccc-0127 | MORSE-E, Program MORSE with Uniform Source for Various Geometry |
| ccc-0588 | MORSE-EMP, Monte-Carlo Neutron and Gamma MultiGroup Transport with Array Geometry, for PC |
| nea-1633 | MOSRA-LIGHT, High Speed 3-D X-Y-Z Nodal Diffusion Code for Vector Computers |
| nea-0933 | MULTI-KENO, Criticality Safety Analysis by Monte-Carlo |
| nea-0035 | MUPO, Critical 43 Group Spectra Calculation for Homogeneous Reactor |
| iaea0890 | MURLI, 1-D Flux, Reaction Rate in Cylindrical Geometry Thermal Reactor Lattice by Transport |
| iaea0892 | MURLI-CLUSTER, Lattice Calculation of PHWR with Rod Clustered Fuel |
| nea-1673 | MVP/GMVP II, MC Codes for Neutron & Photon Transport Calc. based on Continuous Energy and Multigroup Methods |
| iaea1173 | NEHEX-3D, 3-D Neutron Diffusion for Fast Reactors and WWER in Hexagonal Geometry |
| ccc-0641 | NESTLE, Few-Group Neutron Diffusion for Steady-State and Transient Problems by Nodal Expansion Method (NEM) |
| psr-0355 | NJOY-94, General ENDF/B Processing System for Reactor Design Problems |
| psr-0171 | NJOY91, General ENDF/B Processing System for Reactor Design Problems |
| iaea1171 | NOTRAN/3D, 3-D Neutron Transport in X-Y-Z Geometry by Discrete Nodal Transport Method |
| nea-1591 | OMEGA, Subcritical and Critical Neutron Transport in General 3-D Geometry by Monte-Carlo |
| ccc-0266 | ONETRAN, 1-D Transport in Planar, Cylindrical, Spherical Geom. for Homogeneous, Inhomogeneous Probl., Anisotropic Source |
| nea-1324 | OREST, LWR Burnup Simulation Using Program HAMMER and ORIGEN |
| nea-0702 | PALLAS-2DY, 2-D Neutron Transport, Gamma Transport with Anisotropic Scattering for Fixed Source |
| ccc-0760 | PARTISN 5.97, 1-D, 2-D, 3-D Time-Dependent, Multigroup Deterministic Parallel Neutral Particle Transport Code |
| nea-1238 | PASC-1, Petten AMPX-II/SCALE-3 Code System for Reactor Neutronics Calculation |
| nea-0464 | PN, 1-D, 2-D, 3-D MultiGroup Neutron Transport |
| nea-0181 | RADIFLUX, Bessel Function Fit of Radial Flux Data in Cylindrical Reactor |
| nesc0631 | RAFFLE-2/MOD2, 3-D Steady-State Monte-Carlo Neutron Transport, Cell Averaged Scattering Cross-Sections Calculation |
| ccc-0708 | REBUS-PC 1.4, Code System for Analysis of Research Reactor Fuel Cycles |
| nea-0262 | REFLOS, Fuel Loading and Cost from Burnup and Heavy Atomic Mass Flow Calculation in HWR |
| iaea0929 | RICANT, Neutron Transport in X-Y Geometry for Isotropic Scattering |
| nea-0598 | RSYST, Modular System for Reactor Core and Shielding Problems |
| nea-1779 | SAGEP-FR, Sensitivity Analysis of Fast Reactor Parameters |
| ccc-0785 | SCALE 6.1.2, Modular system for criticality, shielding, source term, fuel depletion/decay, inventories, reactor physics |
| nea-1840 | SERPENT 1.1.7, 3-D continuous-energy Monte Carlo reactor physics burnup calculation, lattice physics applications |
| nea-0370 | SHADOK-3-6, Transport Equation with Anisotropic Diffusion in P1 Approximation for Spherical and Cylindrical Geometry |
| nea-0319 | SIMPLED-4, 1-D Neutron Diffusion in Spherical, Cylindrical, Planar Geometry from JAERI Fast-Set and ABBN |
| nea-0905 | SIXTUS-2, 2-D MultiGroup Diffusion in Hexagonal Geometry with Intranodal Solution |
| nea-1426 | SIXTUS-3, 3-D Nodal Neutron Diffusion Criticality in Hexagonal Geometry |
| nea-1081 | SLAROM, Neutron Flux Distribution and Spectra in Lattice Cell |
| nea-0430 | SNAP, MultiGroup 3-D Neutron Diffusion in X-Z, R-Theta-Z, Hexagonal-Z, Triangular-Z Geometry |
| nea-1826 | SOLTRAN, solving multi-dimensional simplified P2 transport and diffusion problems of hexagonal geometry in fast reactors |
| nea-0414 | SQUID-360, 2-D Neutron Diffusion in X-Y and R-Z Geometry with Criticality Search and Constant Neutron Source |
| nea-0703 | STEADY-ACE, 3-D Neutronics and Multichannel Thermohydraulics Analysis of BWR |
| ccc-0248 | SWAN-PPL, Fusion Reactor 1-D Particle Transport Optimization |
| ccc-0204 | SWANLAKE, Cross-Sections Sensitivity Analysis for 1-D Discrete Ordinate Calculation |
| nesc0713 | SYN-3D, 2-D and 3-D Neutron Diffusion Static Eigenvalues, Single Channel Spatial Flux Synthesis |
| iaea1383 | SYRCO-1, 1-D, 2 Groups Multi-Zone Interactive Diffusion Code |
| ccc-0638 | TART2005, 3D Coupled Neutron-Photon Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code |
| nesc0558 | TASK, 1-D MultiGroup Diffusion or Transport Theory Reactor Kinetics with Delayed Neutron |
| nea-0997 | THT, 3-D Coarse Mesh LWR Bundle Fluxes and Power with Discontinuity Factors |
| ccc-0759 | TITAN 1.29, A Three-Dimensional Deterministic Radiation Transport Code System |
| ccc-0543 | TORT, 2-D 3-D Discrete Ordinate Neutron and Photon Transport with Deep Penetration |
| nea-1024 | TP2, Calculation of Reactivity and Kinetic Parameter by 2-D Neutron Transport. Perturbation Theory |
| nea-0900 | TPHEX, MultiGroup Neutron Flux in Homogeneous Hexagonal LWR Cells |
| nea-0953 | TRANSHEX, 2-D Thermal Neutron Flux Distribution from EpiThermal Flux in Hexagonal Geometry |
| nea-0117 | TRAWS-4, Axial Flux Distribution for Control Rod Variations |
| ccc-0293 | TRIDENT, 2-D Neutron Transport for Homogeneous and Inhomogeneous Problems in X-Z, R-Z Geometry, Anisotropic Scattering |
| nea-0384 | TRIGON, 2-D Homogeneous and Inhomogeneous Fixed Source Neutron Diffusion for Triangular or Hexagonal Mesh |
| nea-1716 | TRIPOLI-4.3.3 & 4.4, Coupled Neutron, Photon, Electron, Positron 3-D, Time Dependent Monte-Carlo, Transport Calculation |
| nea-1086 | TRISTAN, 3-D fixed source radiation transport |
| nea-1087 | TRITAC, 3-D Transport by Discrete Ordinate Method in X-Y-Z Geometry |
| nea-0415 | TRITON, 3-D Multi-Region Neutron Diffusion Burnup with Criticality Search |
| nea-0471 | TVEDIM, 2-D Homogeneous and Inhomogeneous Neutron Diffusion for X-Y, R-Z, R-Theta Geometry |
| ccc-0547 | TWODANT-SYS, DANTSYS3.0, 1-D, 2-D, 3-D MultiGroup Discrete Ordinate Method Transport |
| nesc0358 | TWOTRAN-2, 2-D MultiGroup Transport in X-Y, R-Z, R-Theta Geometry with Anisotropic Scattering |
| ccc-0195 | TWOTRAN-GG-FC, General Geometry 2-D Transport with 1st Collision Source Calculation |
| ccc-0195 | TWOTRAN-GG-VW, General Geometry 2-D Transport, Variable-Weight Diamond Difference |
| ccc-0613 | VALE-1.1, 2-D, 3-D MultiGroup Neutron Diffusion for Triangular Problems |
| nesc0264 | VARI-QUIR-3, 2-D MultiGroup Steady-State Neutron Diffusion in X-Y R-Z or R-Theta Geometry |
| uscd1239 | VENTEASY, Criticality Search for a Desired Keffective by Adjusting Dimensions, Nuclide Concentrations, or Buckling |
| ccc-0654 | VENTURE-PC 1.1, Reactor Analysis System with Sensitivity and Burnup |
| nea-1856 | VESTA 2.1.5, Monte Carlo depletion interface code and AURORA 1.0.0, Depletion analysis tool |
| ccc-0754 | VIM 5.1, Steady-State 3-D Neutron Transport Using ENDF/B or Multigroup Cross Sections |
| nesc0510 | VIM, 3-D Monte-Carlo Analysis of Fast Critical Assemblies Using Point Cross-Sections |
| iaea0871 | VPI-NECM, Nuclear Engineering Program Collection for College Training |
| nea-0655 | VSOP, Neutron Spectra, 2-D Flux Synthesis, Fuel Management, Thermohydraulics Calculation |
| iaea1237 | ZZ BARC-75, Coupled 50 Neutron-Group 25 Gamma-Group Cross-Section Library for 42 Nuclides |
| iaea0949 | ZZ BROND, Evaluated Neutron Data Library in ENDF-6 Format |
| dlc-0042 | ZZ CLEAR/42B, 126 Neutron-Group, 36 Gamma-Group Coupled Cross-Section in AMPX, CCCC Format, for LMFBR |
| dlc-0011 | ZZ DLC-11 RITTS, 121-Group Coupled Cross-Section for ANISN, DOT, MORSE |
| dlc-0016 | ZZ DLC-16 COBB, 123 Neutron-Group Cross-Section Library from ENDF/B for XSDRN Calculation |
| dlc-0018 | ZZ DLC-18 NAB, 100 Neutron-Group Cross-Section Library of Na and Al for ANISN, DOT, MORSE Neutron Transport |
| dlc-0002 | ZZ DLC-2D/100G, 100 Neutron-Group Cross-Section Library by SUPERTOG Calculation for ANISN, DOT |
| dlc-0006 | ZZ DLC-6 GAMLIB, 99-Group Cross-Section Library by SUPERTOG Calculation from ENDF/B |
| dlc-0187 | ZZ HILO86R, 66 Neutron, 22 Gamma Group Cross-Section for 400 MeV Neutron, 20 MeV Gamma |
| dlc-0168 | ZZ LA100, ENDF Format Data Library for Neutron and Protons Up to 100 MeV |
| dlc-0054 | ZZ LAFPX-V, Multigroup Fission Product Data Library from ENDF/B-V by Program NJOY |
| nesc0532 | ZZ LASL-XSECS, Fast and Thermal Multigroup Cross-Section Library in LANL Transport Format |
| dlc-0040 | ZZ LIB-IV, 50-Group Cross-Section Library in CCCC-III Format from ENDF/B-IV for Fast Reactors |
| nea-1205 | ZZ MATX175/42-JEFF87, 172 Neutron-Group, 42 Gamma-Group MATXS Library in VITAMIN-J Structure |
| dlc-0176 | ZZ MATXS10, 30-Group Neutron, 12-Group Gamma Cross-Sections in MATXS Format from ENDF/B-VI |
| dlc-0177 | ZZ MATXS11, 80-Group Neutron, 24-Group Gamma Cross-Section in MATXS Format from ENDF/B-VI |
| nea-1206 | ZZ MATXS70-JEFF87, 69+1 Group MATXS Library in WIMS BOXER Structure |
| dlc-0076 | ZZ SAILOR, 47 Neutron-20 Gamma-Group Coupled Cross-Section Library from VITAMIN-C by AMPX |
| dlc-0024 | ZZ SINEX, 100 Neutron-Group Neutron Reaction Cross-Section Library from ENDF/B by SUPERTOG for ANISN |
| iaea0865 | ZZ TEMPEST/MUFT, Thermal Neutron and Fast Neutron Multigroup Cross-Section Library for Program LEOPARD |
| nea-1264 | ZZ VITAMIN J/COVA, Covariance Matrix Data Library for Uncertainty Analysis |
| dlc-0041 | ZZ VITAMIN-C/B, 171 Neutron-Group, 36 Gamma-Group Coupled Cross-Section for Fusion, LMFBR Calculations |
| nea-1207 | ZZ WIMS-LIB/JEF87, 69+1 Group WIMS-D Library from JEF-1 |